This invention relates to a method for immobilizing radioactive wastes for permanent disposal. More particularly, the invention relates to a method of immobilizing mixed waste chloride salts containing radionuclides and other hazardous materials for permanent disposal.
The recovery of fissionable materials such as uranium and plutonium from spent nuclear reactor fuels can be carried out by an electrorefining method using electrochemical cells of the type described in U.S. Pat. Nos. 4,596,647 and 2,951,793, as well as U.S. Pat. No. 4,880,506. It is the electrorefining method which is being developed for the reprocessing of spent nuclear fuel. In a typical electrorefining cell, an electrolyte consisting of a molten eutectic salt mixture such as KCl and LiCl is used to transport the metal or metals to be purified between electrode solutions. When used to reprocess spent nuclear reactor fuels, the salt mixture becomes contaminated with radionuclides, such as cesium.sup.-137 and strontium.sup.-90, hazardous metals such as barium and other species such as sodium and iodine.sup.-129 and eventually is no longer suitable for use in the electrorefining cell.
Ideally the salt would be decontaminated by removing the heat producing radionuclides, primarily cesium and strontium, and any other metals, e.g. sodium, which could potentially interfere in the operation of the electrorefiner and the purified salt would be recycled back to the electrorefiner. However, the separation of cesium and strontium chloride from the salt is difficult, and since they are large heat producers it would be necessary to dilute them in another matrix material and/or cool them before they could be stored. It is therefore more practical to dispose of the cesium and strontium and any other radionuclides and toxic metal chlorides and iodides along with a portion of the salt matrix. The waste salt containing the cesium and strontium is a high level waste (HLW), and as such must be disposed of in the geologic repository for HLW. This requires that the waste form be leach resistant to prevent an uncontrolled release of the radionuclides and other hazardous chemicals such as barium into the groundwater. Since waste salts are chlorides and are very water soluble, a method for encapsulating and immobilizing the waste salt must be identified.
One problem with developing a waste storage medium is that the waste salt consists primarily of chloride salts of the alkali metals and as such is not readily amenable to treatment using procedures and techniques developed for immobilizing the cesium and strontium in other nuclear waste streams. For instance, it has been taught that the chloride salts cannot be added directly to glass-forming compounds and processed to yield a leach-resistant glass since glasses containing halide ions are relatively water soluble, see U.S. Statutory Invention Registration H1,227, published Sep. 7, 1993. Therefore, it was thought that for immobilization in a glass matrix the waste chloride salts must be converted into oxides or other chemical forms compatible with the glass-making process.
However, conversion processes are expensive and time-consuming and raise environmental concerns about the off-gases produced by the processes. A mortar matrix has also been considered as a possible waste form for the waste chloride salt. A special mortar was developed to incorporate lithium, potassium, cesium and strontium chloride salts into its structure and thereby immobilize them. However, when irradiated, the water in the mortar was radiolyzed and large quantities of hydrogen gas were generated.
A new matrix for immobilizing waste chloride salts was therefore needed, and Invention Disclosure H1,227 addressed this problem by disclosing special zeolites which can be treated with molten salts. When some zeolites are treated with molten salts, salt molecules penetrate the cavities and channels of the zeolite and are then said to be occluded. Occluded molecules provide a transfer medium for ion exchange between the cations in the zeolite and those in the bulk salt. A zeolite which has a high selectivity for cesium, strontium and barium would be a promising candidate for an immobilization matrix.
U.S. Pat. No. 5,340,506 which issued Aug. 23, 1994 also addressed the problem by chemically reacting mixtures of NaOH, Al.sub.2 O.sub.3, SiO.sub.2 to form a sodalite intermediate. Further processing produced a sodalite product with radionuclides and hazardous material contained in the sodalite.
As stated in the '506 patent, an advantage of the process of invention registration H1,227 was in the use of certain zeolites to occlude and immobilize waste radioactive chloride salt. Contact between the zeolite (for example zeolite A or mixtures of chabazite and erionite-type zeolites or mixtures thereof) in the sodium, potassium or lithium form and the molten salt resulted in ion exchange between the radionuclides cesium and strontium and the hazardous material barium in the salt and the sodium, potassium, lithium in the zeolite and the occlusion of up to about 25% by weight of the salt within the molecular cavities of the zeolite.
One of the problems inherent in the method disclosed in invention registration H1,227 is that the resultant material is not suitable for storage as a long term waste because it is not a monolithic solid.
Although the use of synthetic naturally occurring minerals to store radioactive ions have been studied, as for instance in U.S. Pat. No. 4,808,318, which describes the use of a modified phlogopite to recover cesium ions from waste solutions and the advances that were set forth in the aforementioned '506 patent there is still needed a method of immobilizing mixtures of salts, particularly chloride salts containing radionuclides and other hazardous wastes so that the highly soluble salts can be safely stored for long periods of time in HLW stored facilities without presenting a hazard to the environment.